ML070740137

Biennial 50.59 Evaluation Report for Period Covering January 1, 2005 to December 31, 2006

ML070740137
Person / Time
Site:	Wolf Creek
Issue date:	03/09/2007
From:	Moles K Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 07-0029
Download: ML070740137 (9)

Text

WSLF CREEK rNUCLEAR OPERATING CORPORATION Kevin J. Moles March 9, 2007 Manager Regulatory Affairs RA 07-0029 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Biennial 50.59 Evaluation Report Gentlemen:

This letter transmits the Biennial 50.59 Evaluation Report for Wolf Creek Generating Station (WCGS), which is being submitted pursuant to 10 CFR 50.59(d)(2). The attachment provides the WCGS Biennial 50.59 Evaluation Report including a summary of the evaluation results.

This report covers the period from January 1, 2005, to December 31, 2006, and contains a summary of 50.59 evaluations performed during this period that were approved by the WCGS onsite review committee.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4126, or Diane Hooper at (620) 364-4041.

Sincerely, KJM/rlt Attachment cc: J. N. Donohew (NRC), w/a V. G. Gaddy (NRC), w/a B. S. Mallett (NRC), w/a Senior Resident Inspector (NRC), w/a

--T--C,-4-7 P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET

Attachment to RA 07-0029 Page 1 of 8 WOLF CREEK NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Docket No.: 50-482 Facility Operating License No.: NPF-42 BIENNIAL 50.59 EVALUATION REPORT Report No.: 20 Reporting Period: January 1, 2005 through December 31, 2006

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Attachment to RA 07-0029 Page 2 of 8

SUMMARY

This report provides a briefý description of changes, test, and experiments performed at Wolf Creek Generating Station (WCGS) and evaluated pursuant to 10 CFR 50.59(c)(1). This report includes summaries of the associated 50.59 evaluations that were reviewed and found to be acceptable by the Plant Safety Review Committee (PSRC) for the period beginning January 1, 2005 and ending December 31, 2006. This report is submitted in accordance with the requirements of 10 CFR 50.59(d)(2).

On the basis of these evaluation of changes:

• There is less than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated Final Safety Analysis Report (USAR).

• There is less than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the USAR.

" There is less than a minimal increase in the consequences of an accident previously evaluated in the USAR.

" There is less than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the USAR.

" There is no possibility for an accident of a different type than any previously evaluated in the USAR being created.

" There is no possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the USAR being created.

" There is no result in a design basis limit for a fission product barrier as described in the USAR being exceeded or altered.

• There is no result in a departure from a method of evaluation described in the USAR used in establishing the design bases or in the safety analyses.

Therefore, all items contained within this report have been determined not to require a license amendment.

Attachment to RA 07-0029 Page 3 of 8 Evaluation Number: 59 2004-0002 Revision: 0

Title:

Risk Informed High Energy Line Break (RI-HELB) Basis Document Activity

Description:

Wolf Creek is implementing a Risk Informed High Energy Line Break (RI-HELB) program, as outlined in EPRI Report 1006937, "Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER) Programs" Rev. 0-A, as an alternative to the criteria for augmented inservice inspection of Branch Technical Position MEB 3-1 as currently stated in the Updated Final Safety Analysis Report (USAR) section 3.6. USAR 3.6 provides criteria for postulating piping breaks. In particular, section 3.6 also defines the requirements that need to be met in order to not postulate piping breaks. One of the criterion involves defining the number of augmented piping inspections that need to be performed on the high energy piping in containment penetration areas. (These are called "No break zones".) These USAR criteria are consistent with Standard Review Plan (section 3.6) criteria.

50.59 Evaluation:

The proposed activity implements an NRC approved methodology as an alternative to the existing USAR requirements. All terms and conditions as stipulated in the NRC SER are met by this proposed activity. Since this proposed activity only affects a "method of evaluation," only question 8 was evaluated.

Attachment to RA 07-0029 Page 4 of 8 Evaluation Number: 59 2005-0004 Revision: 0

Title:

WCGS Rod Withdrawal a't Power Event Safety Analysis Activity

Description:

The Rod Withdrawal at Power event is evaluated by re-analysis to demonstrate that the reactivity and plant control systems are sufficient to prevent Departure from Nucleate Boiling and consequent fuel damage, Over-pressurization and consequent Pressure Boundary Failure, Pressurizer Over-fill and consequent progression of the accident sequence resulting from an uncontrolled rod withdrawal during power operation.

50.59 Evaluation:

This evaluation is performed using RETRAN-3D operating in RETRAN-02 Mode.

In general, this evaluation follows the Westinghouse Safety Analysis Standard and is consistent with the previous evaluations of this event at Wolf Creek as well as the event as described in the USAR. However, there are several changes made in this evaluation that require additional consideration.

1. The Positive Flux Rate Trip is credited for protection

2. RETRAN-3D mod 4.1 in the RETRAN-02 mode was used as the analysis code, instead of RETRAN-02.

3. The Steam Generator Water Level Low-Low Level Trip is no longer credited for protection

4. A Feedwater Control system is included in the model

5. The End of Life Moderator Density Coefficient is increased from 0.47 to 0.50 dK/gm/cc

6. A Limiting condition was sought rather than use of predetermined conditions These changes are considered a change in analysis methodology. All changes were either a result of previously approved changes or were shown to provide conservative or essentially equivalent results.

Attachment to RA 07-0029 Page 5 of 8 Evaluation Number: 59 2005-0005 Revision: 0

Title:

Tube Plugging Criteria for Safety Related Room Coolers Activity

Description:

Cooling loads for the safety related room coolers have increased. These coolers provide cooling to Centrifugal Charging Pump, Safety Injection Pump (SIP),

Containment Spray Pump, Component Cooling Water Pump (CCWP), and Spent Fuel Pool Cooling Pump (SFPCP) rooms. Also, new tube plugging criteria have been established for the SIP, CCWP, Auxiliary Feedwater Pump, and Residual Heat Removal Pump room coolers.

50.59 Evaluation:

The CCWP room-cooling load is less than the cooling capacity of the room coolers. Credit was taken for the available heat sink.

The cooling load for the SFPCP room cooler has increased above its cooling capacity. To compensate, the maximum allowed room temperature for the SFPCP room was increased to 133 OF following a loss of cooling event. The safety related equipment located in the SFPCP rooms was reviewed and determined to be capable of performing its design function at temperatures as high as 140 OF.

With the recommended tubes plugged, these room coolers are capable of maintaining the design ambient temperature of 122 IF in these rooms. However, the spent fuel pool pump room temperature is increased to 133 OF.

All room coolers identified have been shown to provide the design function, i.e.,

provide a suitable environment for the operability of the safety-related equipment located in these rooms.

Attachment to RA 07-0029 Page 6 of 8 Evaluation Number: 59 2005-0006 Revision: 0

Title:

Spent Fuel Pool Storage )Expansion Activity

Description:

The requirement to have one train of the Fuel Building Emergency Exhaust System (EES) in operation during fuel handling~operations will be removed. The change is being made because there is a concern that EES operation during normal fuel movement could adversely affect the assumptions used in the ventilation re-analysis.

Fuel assemblies will be allowed to be stored below the travel path of the spent fuel pool gates during gate movement. The change is being made to increase the available number of storage locations during gate movement.

Two open slots exist below the spent fuel pool gate storage locations. These slots could allow the possibility of a fuel assembly to be accidentally inserted into the slot. The possible mis-positioning of a fuel assembly in these slots is adverse to the UFSAR described function of the fuel racks.

50.59 Evaluation:

Calculation Change Notice AN-99-013-000-CN001 addressed the radiological consequences of not having EES running prior to a fuel handling accident occurring. A conservative one-minute time period of unfiltered release was assumed in the dose calculation to account for the delayed initiation of EES. The increase in dose is less than 10% of the difference between the current calculated value and the regulatory guideline value. Therefore, this change does not result in more than a minimal increase in the consequences of an accident previously evaluated in the USAR.

An analysis was done on a postulated gate drop event for the new spent fuel racks in its storage location. A report conducted by Holtec addressed the heavy load issue, pertaining to the spent fuel pool gates. The gate drop accident analysis concluded that no damage to fuel would occur. The fuel handling accident outside containment would envelope this event. Deformation of the upper portion of the rack provides sufficient energy dissipation to preclude penetration to the depth of the top of the fuel assembly.

Though administrative controls are in place to preclude fuel assemblies from being place in the two open slots, criticality analysis evaluated the possibility of a fresh fuel assembly being dropped into one of the two open slots. The case is bounded by the more severe fuel mis-positioning case.

Attachment to RA 07-0029 Page 7 of 8 Evaluation Number: 59 2006-0901 Revision: 0

Title:

Radiological Consequences of a Fuel Handling Accident Activity

Description:

The radiological consequences analysis of a fuel handling accident are revised to allowed for an earlier start of fuel movement following reactor shutdown, specifically, the radiological consequences of a fuel handling accident at 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> instead of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.

50.59 Evaluation:

This change, reducing the required decay time prior to the start of core alteration or movement of irradiated fuel, affects the assumed initial activities in the radiological consequences analysis of a fuel handling accident as described in USAR Section 15.7.4. Calculation AN-04-015 Revision 1, based on the updated initial activities, showed that the radiological consequences of a postulated fuel handling accident remain well within the guideline values of 10CFR100. Also, a conservative one-minute time period of unfiltered release was assumed in the doses calculation to account for the delayed initiation of the emergency ventilation system. The increases are less than 10 percent of the difference between the current calculated dose value and the regulatory guideline value.

Therefore, this change does not result in more than a minimal increase in the consequences of an accident previously evaluated in the USAR.

Attachment to RA 07-0029 Page 8 of 8 Evaluation Number: 59 2006-0002 Revision: 0

Title:

Operations Procedure, GEN 00-004 Revision 54, Power Operation Activity

Description:

The current version of the Operations procedure GEN 00-004, 'Power Operation', uses the Pressurizer Water Level Control System (PWLC) to decrease pressurizer level from the full load 57% span to the 27% span lower limit during plant shutdown from 100% to 0% power. Per procedure, manual actions to restore the pressurizer level are then undertaken to accommodate shrinkage of the Reactor Coolant System during cooldown.

The proposed change to procedure GEN 00-004 will allow the pressurizer level to be maintained at full load during plant shutdown from 100% to 0% power.

This effectively removes the reliance on the automatic operation of the PWLC system and then eliminates the subsequent restoration of the pressurizer water level.

50.59 Evaluation:

This evaluation concludes that the proposed change, with the pressurizer level being maintained at the full load level during plant shutdown from 100% power to 0% power prior to cooldown, is acceptable. Accident analyses for WCGS, assuming a typical 5% span pressurizer level uncertainty, demonstrate no pressurizer overfill at full power. Applying the restrictive criterion that water discharge through the pressurizer safety valves is precluded by demonstrating no pressurizer overfill, is a further conservatism. Therefore, current accident analyses bound postulated accident scenarios based upon maintaining the higher than programmed level in the pressurizer during plant shutdown from 100% to 0% power and the subsequent refill period. The effects on accidents and malfunctions previously evaluated in the USAR are negligible.